Journal of Nuclear Fuel Cycle and Waste Technology

Maintaining the structural integrity of spent nuclear fuel (SNF) cladding under transport and handling conditions is essential for safe and economic management of SNF. Reoriented hydrides significantly reduce the fracture resistance of cladding under pinch loads, necessitating an investigation of its fracture behavior. In this study, a simulation model was developed for cladding under ring compression test (RCT) using ductile damage model. This model enables the simulation of crack initiation and propagation under various stress triaxiality conditions. However, the calibration of the fracture parameter of the developed model is challenging due to the lack of experimental data and complexity of the parameter space. To address this, a metamodel-based optimization framework was proposed to calibrate the fracture parameters of the Zr/hydride interface, which exhibits the lowest load resistance, using RCT data. Two optimization algorithms - global search algorithm (GSA) and a genetic algorithm (GA) - were employed and their execution time, accuracy, and practicality were compared. Both algorithms produced nearly identical solutions. The optimized parameters were validated against experimental RCT data and demonstrated high prediction accuracy for crack initiation load and displacement during RCT.

For a fundamental assessment of spent nuclear fuel integrity, the fatigue behavior of cladding tubes was investigated by evaluating the fatigue strength reduction factor (K_f) due to wear and oxidation-induced degradation. In this study, as wear geometry significantly influences fatigue life cycles leading to fracture, it was determined based on the contact conditions observed in commercial fuel designs. Wear and oxidation-affected cladding tubes were fabricated, along with reference specimens without wear, for comparative analysis. Fatigue tests were conducted by applying and releasing cyclic high internal rod pressures (<102 MPa) at 4 Hz to simulate operational conditions. The obtained average K_f for wear specimens was 1.04, based on an effective stress analysis relative to the unworn reference tube. However, oxidation led to a significantly greater reduction in fatigue life with K_f values exceeding 1.5, indicating more dominant role in fatigue degradation. It is noteworthy that oxidation‑induced metal wastage caused a significantly greater reduction in fatigue life than wear, thereby emphasizing the need for stringent corrosion control in dry storage environment.

Two-dimensional MXenes (Ti3C2Tx) have received extensive attention due to their easy availability, hydrophilic behavior, and tunable chemical properties; however, their permeability-selectivity trade-off and self-stacking property limits their application in water treatment. In this study, three-dimensional Ti3C2Tx films with high adsorption capacity, permeability, and high selectivity were successfully fabricated using a simple vacuum-assisted filtration method for the removal of radionuclides (Cs+, Sr2+, and Co2+) from aqueous solutions. Various characterization results suggested that radionuclides tended to coordinate with heterogeneity of the sites (containing hydroxyl, fluorine, and oxygen groups) on Ti3C2Tx, as well as the multilayer adsorption properties of the Ti3C2Tx films. The synthesized Ti3C2Tx film demonstrated maximum removal efficiencies of 99.50%, 99.54%, and 99.8% for Cs+, Sr2+ and Co2+, respectively, at pH 7 and 25℃. The adsorption behavior and process analysis showed that the adsorption mechanism primarily involves electrostatic and chemical interactions on the Ti3C2Tx film. Furthermore, the films not only sequestrated radionuclides from the aqueous solution, but also selectively separated target radionuclides by adjusting the Ti3C2Tx interlayer size. Therefore, the newly developed Ti3C2Tx films demonstrated great potential for the removal of radionuclides from aqueous solutions.

This study investigated the unconfined compression behavior of unsaturated compacted Ca-bentonite (dry density = 1.6 Mg/m3) after exposure to 0.02 M Ca(OH)2 solution and elevated temperatures (25 °C and 150 °C) for up to 28 days. Unsaturated Ca-bentonite demonstrated brittle failure behavior regardless of the experimental conditions. At 25 °C, the unconfined compression strength (qu) of Ca-bentonite specimens mixed with either deionized (DI) water (pH 6) or Ca(OH)2 solution (pH 12) was statistically unchanged. However, at 150 °C, qu values increased significantly compared to 25 °C. A similar trend was observed for Young’s modulus, indicating increased stiffness at higher temperatures. The enhancement is attributed primarily to reduced water content and corresponding increases in suction rather than direct mineralogical alteration, as confirmed by X-ray diffraction analysis. This study provides the early-stage mechanical response of unsaturated Ca-bentonite under repository-relevant thermal and alkaline conditions, offering understanding into its performance before saturation is achieved.

A method for predicting the level of radioactive waste during the preliminary decommissioning stage of a nuclear reactor is presented, utilizing both Monte Carlo and deterministic codes simultaneously. An autonomous simulation code, developed in Python, was created to estimate radioactive waste during the decommissioning of both PWR and CANDU reactors. The materials considered for activation analysis include only concrete, stainless steel, and vessel steel. Neutron flux is calculated using MCNP, a Monte Carlo-based code, while SCALE-ORIGEN, which is based on the matrix exponential method, is used to perform the activation analysis. The developed RSAM (Reactor Structures Activation Module) serves as a bridge between these two codes, enabling the automatic generation of activation analysis input files and the interpretation of results. Additionally, RSAM can autonomously conduct activation sensitivity analysis based on material data from NUREG-3474. To verify RSAM’s performance, activation analyses were conducted on the structural components of CANDU and OPR-type reactors. The developed Module effectiveness was demonstrated through case studies on CANDU and OPR reactor (→ PWR and PHWR), indicating its broad applicability and potential as a valuable tool for nuclear power plant decommissioning projects. 

Spent ion exchange resins (SIERs) are commonly immobilized in cement, necessitating the reliable determination of radionuclides in cement-based waste forms. This study developed a simple and practical method for quantifying 129I, a key radionuclide in low- and intermediate-level radioactive waste (LILW) disposal. Investigation into iodine behavior during leaching, extraction, and precipitation revealed that iodate (IO3-) readily adsorbs onto cement minerals, reducing chemical recovery in extraction-based methods. To mitigate this, we implemented an in-situ conversion of IO3- to I2, improving chemical yields from 58% to 93% in fine cement samples ( < 0.05 mm). The optimized method was further applied to cement containing anion exchange resins, where increasing resin content led to a decrease in chemical yield, reaching 51% at 10wt% resin loading, close to the practical limit in ordinary Portland cement (OPC) waste forms. This decline was likely attributed to the re-adsorption of iodide (I-) onto the resins during the conversion of IO3- to I2. Despite this, measured 129I activities closely matched expected values, demonstrating its reliability for routine analysis of cement-based radioactive waste.

With an emphasis on climate change and long-term perspectives, this study discusses the impact of rainfall patterns on the safety and integrity of near-surface radioactive waste disposal facilities in South Korea. Future rainfall and infiltration scenarios are prepared by using historical rainfall data from the Ulsan district, and the impact of rainfall patterns is modeled with COMSOL Multiphysics and GoldSim. Infiltration patterns do not significantly affect the total annual dose. However, they do have a minor impact on the total annual dose from 129I and 90Sr, contributing a smaller dose during the 1000-year simulation period. Radionuclides such as 3H, 99Tc, and 14C are the primary contributors to the total annual dose, which results from the radionuclide concentrations in the saturated rock zone, regardless of the assumed cases of inputs such as waste zone permeability and saturated zone groundwater flow rate. This study provides essential insights and recommendations for the safe management and design of multi-layered cover system in radioactive waste disposal facilities, considering the evolution of long-term climatic conditions.

This study introduces a strain limit-based evaluation method that considers stress triaxiality to assess the structural integrity of radioactive waste transport containers. Unlike conventional stress-based approaches, this method provides a more precise assessment of localized plastic deformation and damage. By incorporating strain-based failure criteria, it enables a realistic evaluation of impact-induced deformation under complex stress states. A post-processing program is developed to intuitively verify strain safety margins, improving assessment accuracy. Additionally, sub-modeling techniques refined the strain distribution analysis in damage-concentrated areas, which could not be captured in full-scale simulations. This enhances the reliability of impact evaluations. Based on these procedures, a drop safety verification process was established for the freight container used in low- and intermediate-level waste transport. The results demonstrated the feasibility of strain limit-based evaluation in practical applications. This approach improves the structural assessment of transport containers and contributes to the advancement of design and certification processes.

The sorption of Europium(III) (Eu(III)) on MX-80 bentonite and granite was systematically studied as a function of ionic strength (0.05 – 1 mol·kgw−1) and pHm (4–9) in Ca-Na-Cl solutions by batch experiments and 2-site protolysis no electrostatics surface complexation and cation exchange (2SPNE/CE) model. Eu(III) has often been used as a chemical analogue of plutonium(III) which is an important element of interest for the safety assessment of a deep geological repository (DGR) for nuclear waste. It was found that the sorption of Eu(III) on MX-80 and granite was dependent on pHm. The sorption of Eu(III) on MX-80 was dependent on ionic strength at pHm ≤ 6. The effect of ionic strength on the sorption of Eu(III) on granite was negligible. The 2SPNE/CE model simulation suggested that the ion exchange reaction might be the main sorption mechanism at low pHm, while the surface complexation reactions might be the dominant sorption mechanism at higher pHm for the sorption of Eu(III) on both MX-80 and granite.

Deep geological disposal facilities require early regulatory involvement to ensure safety, transparency, and efficiency throughout their long development process. Leading countries, such as Finland, Sweden, France, and the UK, actively engage regulatory bodies from the early stages, following international guidelines set by the IAEA and OECD/NEA. However, in Korea, the current regulatory framework lacks a legal basis for pre-licensing reviews, highlighting the need for institutional reforms and clear agreements between regulators and operators. A phased regulatory approach which covers pre-site selection, site selection, and post-selection stages can enhance project stability and minimize risks. To successfully implement early regulatory involvement, Korea must revise its legal framework, develop strategic roadmaps, and foster transparent communication with stakeholders.